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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Port Gamble S'Klallam Tribe Strategic Energy Plan

The Port Gamble S'Klallam Tribe's Strategic Energy Plan (SEP) details the outcomes and efforts of a 1.5-year Energy Technology Innovation Partnership Program (ETIPP) project. This collaboration involved the Port Gamble S'Klallam Tribe, Spark Northwest, the Pacific Northwest National Laboratory (PNNL), and the National Renewable Energy Laboratory. Project partners supported the Tribe in identifying key energy research focus areas and the Tribe's energy vision. As the technical lead on the project, PNNL led a series of research initiatives to evaluate potential pathways for the Tribe to achieve its energy goals, subsequently providing associated recommendations. The SEP offers a comprehensive analysis encompassing several critical aspects including: -The Tribe's current energy landscape -Availability of renewable resources -Potential microgrid designs, developed by Säzän Group -Critical infrastructure analysis -Electric vehicle integration -An energy workforce development framework, led by Spark Northwest The analysis results come with recommended next steps, categorized by implementation timeframe (short-term, medium-term, long-term) and priority.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2022

Analysis of SCALE Criticality and Sensitivity Calculations for Reflected HEU Cylinders [Abstract]

The SCALE code package offers several nuclear data libraries to support Monte Carlo (MC) transport, as well as MC-based derivation of $\kappa$ eff sensitivity and uncertainty (S/U) data. The CSAS sequence using the KENO MC code can utilize continuous-energy (CE) cross sections, or pre-generated multigroup (MG) cross section libraries. The use of MG libraries introduces bias into calculations in exchange for faster transport solutions. The TSUNAMI-3D sequence also utilizes KENO MC calculations. TSUNAMI-3D has two CE calculational methods: the Iterated Fission Probability (IFP) method, and the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Tracklength importance CHaracterization (CLUTCH) method. Previous work has shown poor agreement between CLUTCH and confirmatory direct perturbation calculations in specific applications, e.g., fissionable and polyethylene reflectors

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Analysis of NCERC Critical Experiments with ENDF/B Nuclear Data Libraries

Nuclear data (ND) libraries are the backbone of the nuclear industry, as they are the collections of tabulated probabilities that define sub-atomic particle interactions with matter. In the areas of criticality safety and experiments, no evaluated nuclear data files (ENDF) are more important than those containing neutron cross section data. It is upon these files, and accompanying radiation transport codes, that practitioners are enabled to safely design both subcritical and critical systems. Likewise, in a symbiotic fashion, it is the same critical assemblies which are used primarily to validate that the cross sections are correct. The ENDF/B library, the United States’ national library maintained by the National Nuclear Data Center (NNDC) at Brookhaven National Laboratory (BNL), is soon releasing a new version, ENDF/B-VIII.1. Prior to the official release, several beta versions of the library were prepared and tested in simulation suites. The work presented here are results comparing the newest ENDF/B beta release (ENDF/B-VIII.1b3) and ENDF/B-VIII.0 with recent experiments done at the National Criticality Experiments Research Center (NCERC) as well as correlated experiments from the Los Alamos Critical Experiments Facility (LACEF). These NCERC and LACEF experiments were performed in part to provide validation for various cross sections that were identified as insufficient in the ENDF/B-VIII.0. In particular, lead, copper, fluorine, and chlorine, as well as the major actinides, were targeted from the last decade of critical experiments.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Preliminary Criticality and Radiation Shielding Analysis for the Storage and Transfer of MARVEL Reactor Spent Nuclear Fuel

This report documents the results of preliminary nuclear criticality and radiation shielding assessments during transfer and dry storage of irradiated Microreactor Applications Research Validation and Evaluation (MARVEL) reactor fuel at Idaho National Laboratory (INL). The assessments focus on transfer casks and storage canisters that are currently in use at INL, which may be compatible with the irradiated MARVEL reactor fuel. The criticality assessments were performed using the radiation transport code MCNP6 with 37 MARVEL reactor fuel elements in various configurations and scenarios. All transfer and storage configurations under dry conditions were below the assumed criticality safety limit of 0.93. Some storage and transfer configurations under wet conditions exceeded the criticality safety limit. This suggests that the appropriate administrative and engineering controls, in addition to reducing the number of MARVEL reactor fuel elements per transfer cask or storage canister, can be expected to ensure criticality safety under all scenarios. The radiation shielding assessments were performed by generating conservative neutron and photon source terms using the ORIGEN module in the SCALE suite of codes. These source spectra were used to estimate the dose equivalent rates using the radiation transport code MCNP6, both on contact and 1 m away from the fuel and transfer casks. The maximum estimated dose equivalent rate of 37 unshielded MARVEL reactor fuel elements on contact is approximately 42000 R/hr. The maximum estimated dose equivalent rates on contact to the ATR transfer cask, HFEF-5 transfer cask, and high load charger were approximately 233 mR/hr, 171 mR/hr, and 201 mR/hr, respectively. This suggests that with the appropriate administrative and engineering controls, all three transfer casks analyzed can be expected to provide sufficient radiation shielding to workers during transfer of irradiated MARVEL reactor fuel. These calculations are performed to support the planning and strategy for the MARVEL project and will demonstrate the technical viability of the different configurations discussed and help identify where engineered or administrative controls may be necessary. A complete criticality safety analysis and radiation shielding analysis, including validation and contingency and accident analysis must be completed by licensed and authorized personnel before any transfer or storage of MARVEL reactor nuclear fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

EVALUATION OF HRA METHODOLOGIES FOR APPLICATION IN SDP WORK

This study critically evaluates human reliability analysis (HRA) methodologies applicable to regulatory probabilistic safety assessment (PSA) model, with a particular focus on their role in supporting the significance determination process (SDP) in nuclear safety assessment. Firstly, three widely utilized HRA methods – IDHEAS-ECA, SPAR-H, and ASEP/THERP – were qualitatively and quantitatively assessed. Qualitative assessments were conducted using attributes from the NEA/CSNI/R(2015)1 report, while quantitative evaluations employed regression and correlation analyses to compare predicted human error probabilities (HEPs) against empirical data. Results reveal distinct strengths, for example, IDHEAS-ECA’s robust predictive accuracy and K-HRA’s alignment with operational practices. In addition, dependency analysis and recovery analysis were critically evaluated. For dependency analysis, the methods’ handling of inter-task dependencies and their impact on HEPs were examined, while recovery analysis highlighted strategies for mitigating failure events. Furthermore, strategies were proposed to evaluate performance-shaping factors under conditions of reduced human performance, such as stress, fatigue, or cognitive overload, addressing specific challenges faced in SDP evaluations. Human errors from KINS’s operational performance information system event reports were evaluated as a case study. This study identifies gaps and provides actionable insights to ensure their validity and applicability in SDP HRA applications. This paper is a part of research conducted by KINS, and it should be noted that this result does not represent the regulatory position of KINS.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2022

Prediction of the energetic particle redistribution by an improved critical gradient model and analysis of the transport threshold

Based on the theory of critical gradient model (CGM) and following the simulation method proposed by Waltz et al., a combination of TGLFEP and EPtran code is employed to predict the energetic particle (EP) transport induced by Alfvén eigenmodes (AEs). To be consistent with the experiment, recent improvements to the simulation method include consideration of threshold evolution and orbit loss due to finite orbit width. The revised CGM is applied to simulate two DIII-D experimental discharges (#142111 and #153071). It well reproduces the experimental profiles with multiple unstable AEs and large-scale EP transport. Discharge #142111 had previously been simulated using a nonlinear MHD-kinetic code MEGA with a transport mechanism based on stochasticity induced by overlapping AE. By comparing the simulated EP profiles, we find that the AE transport threshold is approximated by both the MEGA nonlinear stability threshold and the proposed CGM threshold (error <5% for single n and <17% for multiple n simulation). Both of them are larger than the linear stability threshold of the most unstable AE mode by a quantity of the order of the flux needed to sustain EP transport by the background turbulence. Additionally, we have also applied the improved CGM to simulate the α particle redistribution for a China Fusion Engineering Test Reactor steady state scenario. Because of the clear separation between the AE unstable region and the loss cone, only a moderate α particle loss of ~9.6% is predicted.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2024

CI-MOR Final Report: Analysis and Validation of Critical Infrastructure Models using Model Order Reduction

This report summarizes the research and capabilities developed as part of the project “Analysis and Validation of Critical Infrastructure Models using Model Order Reduction” (CI-MOR) LDRD project. CI-MOR research enables the solution of large, complex optimization models that naturally arise in national security challenges involving critical infrastructures. Specifically, CI-MOR researchers developed methods to (1) rigorously approximate complex, nonlinear optimization formulations, (2) identify alternative near-optimal solutions, (3) accelerate optimization workflows used for complex applications, and (4) rigorously integrate domain knowledge in stochastic-process models. This report provides an overview of the research done in CI-MOR, and we describe application exemplars used to illustrate CI-MOR capabilities. Furthermore, we describe the software developed by CI-MOR that researchers can leverage to analyze new applications.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Critical Material Supply Chain Analysis: Magnetocalorics

Understanding manufacturing cost and supply chain implications of technologies is critical to aid adoption of next-generation energy solutions, such as advanced heating ventilation and air conditioning (HVAC) equipment for building energy efficiency. This report provides insights into the market, regulatory, technology, and cost drivers most impacting the adoption of magnetocaloric refrigeration (MCR) systems in the U.S., with a focus on the rare earth materials requirements, manufacturing costs of key components, and detail on the supply chain of critical materials.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2022

An ontology-based fault generation and fault propagation analysis approach for safety-critical computer systems at the design stage

Abstract Fault propagation analysis is a process used to determine the consequences of faults residing in a computer system. A typical computer system consists of diverse components (e.g., electronic and software components), thus, the faults contained in these components tend to possess diverse characteristics. How to describe and model such diverse faults, and further determine fault propagation through different components are challenging problems to be addressed in the fault propagation analysis. This paper proposes an ontology-based approach, which is an integrated method allowing for the generation, injection, and propagation through inference of diverse faults at an early stage of the design of a computer system. The results generated by the proposed framework can verify system robustness and identify safety and reliability risks with limited design level information. In this paper, we propose an ontological framework and its application to analyze an example safety-critical computer system. The analysis result shows that the proposed framework is capable of inferring fault propagation paths through software and hardware components and is effective in predicting the impact of faults.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2025

Interpretable machine learning models classify minerals via spectroscopy

Developing methods to identify mineral species confidently and rapidly from Raman spectral analysis is critical to numerous fields. Traditionally, analysis relies on pattern matching the Raman spectrum of an unknown dataset with a supporting library of well-characterized spectral data, which may prove difficult for environmental samples that are poorly crystalline or phase mixtures. Here, we developed interpretable machine learning models that can classify uranium minerals by secondary oxyanion chemistry and other physicochemical properties based solely on Raman spectra. This new ML method produces a mineral profile of physical and chemical properties for an unknown sample and can rapidly classify or identify unknown minerals from Raman data, without the need for an exact pattern match in a spectral library. Training models are validated by 1. Strong correlation of high confidence model regions with published spectroscopic assignments and 2. Correct classification of a mineral not present in training data. Training data are from the Compendium of Uranium Raman and Infrared Experimental Spectra and available crystallographic information files within the open-source Smart Spectral Matching scientific framework. Physically meaningful classifier models can rapidly identify key structural and chemical information about unknown uranium minerals and the overall methodology is broadly applicable for mineral phases.

Machine learning↗
DOE OSTI2021

Critical Roles of Information, Analysis, Research, and Operations in the Cyber Realm

PNNL has developed an adaptive cyber integration framework (ACIF) to facilitate the timely sharing of cyber threat information along with the advancement of situational awareness tools to enhance protection against and respond to critical infrastructure cybersecurity threats. The ACIF comprises components implemented iteratively to achieve research and mission goals. The ACIF components include data generation technologies, analytic tools development and maturation, data enrichment and fusion, trust building with stakeholders, investigative research, analytic rigor, production, and dissemination. Each component, its importance to the ACIF, and how they can be adopted and applied across other information-sharing sectors and domains are discussed in this paper.

Cyber Threat Intelligence, Cyber Security, Data En↗
DOE OSTI2022

Hand Calculations for Nuclear Criticality Safety – Primer Revision

The document “ Hand Calculation Methods for Criticality Safety – A Primer, ” was published by the authors (Los Alamos National Laboratory report LA-14244-M) in December 2006 to provide an overview of the most common hand calculation methods used for criticality safety analysis. Hand calculation methods can be used as a starting point for more advanced calculations, and in many circumstances, they can provide sensitivity and perturbation information quicker than using a criticality code, such as KENO or MCNP. This report has been revised to augment existing basis information and example problems for each of the methods discussed: 1-group and modified 1- group diffusion theory, core-density conversions, buckling conversions, limiting surface density method, density analog method, surface density method, and solid angle method. This primer discusses the applicability of the various methods, illustrates how they are used, and provides an interpretation of the example problem results. The example problems provided are simple, practical problems, and the reference data to solve each example problem is provided. This paper will provide an overview of the Primer revision and how it can be useful for new Nuclear Criticality Safety practitioners.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Hand Calculations for Nuclear Criticality Safety — Primer Revision [Abstract]

The document “ Hand Calculation Methods for Criticality Safety –A Primer, ” was published by the authors (Los Alamos National Laboratory report LA-14244-M) in December 2006 to provide an overview of the most common hand calculation methods used for criticality safety analysis. Hand calculation methods can be used as a starting point for more advanced calculations, and in many circumstances, they can provide sensitivity and perturbation information quicker than using a criticality code, such as KENO or MCNP. This report has been revised to augment existing basis information and example problems for each of the methods discussed: 1-group and modified 1-group diffusion theory, core-density conversions, buckling conversions, limiting surface density method, density analog method, surface density method, and solid angle method. This primer discusses the applicability of the various methods, illustrates how they are used, and provides an interpretation of the example problem results. The example problems provided are simple, practical problems, and the reference data to solve each example problem is provided. This paper will provide an overview of the Primer revision and how it can be useful for new Nuclear Criticality Safety practitioners.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗